JP5570481B2 - Reactor core performance calculator - Google Patents

Description

  The present invention relates to a reactor core performance calculation apparatus, and more particularly to a reactor core performance calculation apparatus suitable for monitoring the life of a control rod.

The control rod used in the boiling water reactor has a cross shape in cross section, and has four blades extending in four directions from a tie rod arranged at the axial center. Each blade has a sheath having a U-shaped cross section attached to the tie rod by welding, and a plurality of neutron absorber rods disposed in the sheath. Each neutron absorber rod is filled with B ₄ C, which is a neutron absorber. A handle is attached to the upper end of the tie rod by welding, and the upper end of the sheath of each blade is also welded to the handle. The lower support member is attached to the lower end portion of the tie rod by welding, and the lower end portion of the sheath of each blade is also welded to the lower support member.

The life of such a control rod will be described. The life of the control rod has a nuclear life and a mechanical life, and the control rod reaches the end of its life when one of the lifespans is reached. In general, in the control rod using B ₄ C, the nuclear life is shorter than the mechanical life, so the nuclear life becomes the life of the control rod. B ₄ C contains boron 10 (B10) which is a nuclide having a very large thermal neutron absorption cross section, and this B10 mainly absorbs neutrons. B10 is converted into Li and He that hardly absorbs neutrons by absorption of neutrons. For this reason, in the state where neutrons are irradiated to the control rods arranged in the core, the ability of the control rods to absorb neutrons, that is, the value of the control rods, is increased with the passage of time after the placement of the control rods in the core. It goes down. The nuclear lifetime is the relative of the control rods in any section obtained by dividing the length of the region filled with neutron absorber (for example, B ₄ C) (the effective length of the control rods) into four equal parts in the axial direction of the control rods. This is when the value has decreased by 10%.

By the way, among the neutron absorber rods in the blade, the neutron absorber rod arranged on the tip side of the blade has a larger area facing the water, so that the neutron absorber rod arranged on the tip side generally has a higher neutron irradiation amount. . In a neutron absorber rod arranged at the tip of a blade with a large amount of neutron irradiation, the cladding may be damaged due to swelling before the value of the control rod is reduced by 10%. Therefore, in the life evaluation of the control rod, the B10 consumption rate at which the cladding may be damaged due to swelling is determined conservatively, and when the B10 consumption rate is reached, the neutron absorber rod is further conservatively reached. Suppose that the B ₄ C powder of the inside disappears completely. Therefore, as compared with the case where damage of the cladding tube due to swelling is not considered, the time for the control rod value to decrease by 10% is quickened, and the life of the control rod is considerably shortened. Thus, the point at which the control rod value decreases by 10% is not simply determined by the nuclear reaction between neutron and B10. For this reason, the control rod average B10 consumption rate corresponding to the time when the value of the control rod in consideration of breakage of the cladding tube due to swelling decreases by 10% is set and redefined as the nuclear life.

  A control rod that is expected to reach the end of its life in a certain operating cycle is replaced with a new control rod during the previous periodic inspection. In order to judge the replacement of the control rod, the life of the control rod is monitored in the core performance calculation. Specifically, a monitoring index for monitoring the life is set, and the monitoring index is evaluated by core performance calculation. In the conventional core performance calculation, the monitoring index is the thermal neutron irradiation amount. A limit value of the thermal neutron irradiation amount corresponding to the lifetime is set, and when the thermal neutron irradiation amount as a monitoring index reaches the limit value, it is determined that the control rod has a lifetime.

  The calculation method of the thermal neutron irradiation amount in the conventional core performance calculation will be described. In this core performance calculation, the reactor core state data such as core flow rate, core thermal output, and control rod position, plant data such as measured values of neutron detectors arranged in the core, and pre-calculated for each fuel assembly are calculated. The fuel output for each unit (referred to as a node) obtained by dividing the effective fuel length of the fuel assembly into 24 in the axial direction of the fuel assembly is evaluated using the nuclear constant. In the conventional core performance calculation, the thermal neutron flux at each node is obtained by multiplying this fuel output by the core average power-thermal neutron flux conversion factor evaluated separately to obtain the thermal neutron flux for each node. To obtain the thermal neutron dose of the control rod.

  By the way, recently, a core loaded with a plurality of fuel assemblies in which nuclear fuels having greatly different enrichments are mixed, and a plurality of fuel assemblies in which uranium fuel and mixed oxide fuel (MOX fuel) are mixed. The loaded core has been put into practical use. In general, since the relationship between fuel output and thermal neutron flux varies depending on the nuclear fuel, the core conversion factor for the conventional core average is used for a core loaded with multiple fuel assemblies containing nuclear fuels with greatly different characteristics as described above. The accuracy of the thermal neutron dose determined by the method used is reduced. For this reason, in such a core, it is necessary to conservatively set a thermal neutron dose limit corresponding to the lifetime.

  As a method for accurately evaluating the thermal neutron irradiation amount of the control rod, for example, there is a method described in JP-A No. 2000-162374. In this method, the thermal neutron irradiation amount of the control rod is obtained using an output-thermal neutron flux conversion factor that takes into account the types of four fuel assemblies adjacent to the control rod. In the method described in Japanese Patent Application Laid-Open No. 2000-162374, the difference in output-thermal neutron flux conversion coefficient depending on the fuel type is taken into consideration, so that the thermal neutron irradiation amount on the control rod can be accurately evaluated.

JP 2000-162374 A

  However, as a result of examining the neutron irradiation amount calculation method described in Japanese Patent Application Laid-Open No. 2000-162374, the inventors have found that there are two more problems.

  The first problem is about the evaluation accuracy of the thermal neutron irradiation dose. The power-thermal neutron flux conversion factor varies depending on the void ratio and burnup even if the fuel type is the same. In the above prior art, if the fuel type is the same, the same output-thermal neutron flux conversion factor is used, and the difference in void fraction and burnup cannot be considered. The fuel type is information that classifies the fuel assembly. Not only the type of fuel assembly is classified by the shape of the fuel assembly, but also by the difference in the concentration distribution and the concentration distribution of the flammable poison even if the shape is the same. The fuel assemblies are classified.

  The second problem is a problem regarding the correlation between the thermal neutron irradiation amount and the B10 consumption rate of the control rod average, which is the definition of life. The relationship between the thermal neutron irradiation amount and the control rod average B10 consumption rate differs depending on the fuel type, its void rate, and burnup. That is, even if the thermal neutron irradiation amount is accurately evaluated, the value of the B10 consumption rate of the corresponding control rod average varies depending on the fuel type, the void ratio, and the burnup, so the thermal neutron irradiation amount is good. It is not a monitoring indicator. For this reason, it is necessary to conservatively set a limit value of the thermal neutron irradiation amount corresponding to the life of the control rod. Thus, setting the limit value conservatively is a sufficient measure for safety. However, since it will have an excessive margin for the service life, the control rods are removed from the reactor for replacement in addition to the fact that the replacement time of control rods is advanced and the number of exchange bodies increases, which is uneconomical. Will lead to an increase in radioactive waste.

  An object of the present invention is to provide a core performance calculation apparatus capable of accurately obtaining a monitoring index having a good correlation with a control rod life.

The feature of the present invention that achieves the above-described object is a conversion coefficient for converting the fuel output of the fuel assembly adjacent to the control rod for obtaining the monitoring index into the monitoring index, using the fuel type, void ratio, and burnup as parameters. A conversion coefficient storage device that stores the fitting coefficient of the conversion coefficient, a fitting coefficient stored in the conversion coefficient storage device, and a void fraction and combustion of the fuel assembly input from the three-dimensional core nuclear thermal hydraulic analysis device degrees to calculate the conversion factor by using, based on the fuel output to input from the conversion factor and the three-dimensional core nuclear thermal hydraulic characteristic analyzing apparatus, as a monitoring indicator of the control rod lifetime, neutron absorption amount of the control rods, control rod inside And a monitoring index analyzer for obtaining one of the neutron absorption amount of B10 and the control rod average B10 consumption rate .

Since the monitoring index having a good correlation with the life of the control rod is obtained by using the conversion factor with the fuel type, void ratio and burnup as parameters, the conversion factor from the fuel output to the monitoring index, and the fuel output, the control rod The control rod neutron absorption, the neutron absorption of B10 in the control rod, or the control rod average B10 consumption rate, which is a monitoring index having a good correlation with the lifetime, can be obtained with higher accuracy .

  According to the present invention, a monitoring index having a good correlation with the control rod life can be obtained with high accuracy.

It is a block diagram of the reactor core performance calculation apparatus of Example 1 which is one suitable Example of this invention. It is a perspective view of the control rod used for a boiling water reactor.

  Examples of the present invention will be described below.

  A reactor core performance calculation apparatus according to embodiment 1, which is a preferred embodiment of the present invention, will be described with reference to FIG. The reactor core performance calculation apparatus 1 of this embodiment includes a plant data input device 2, a nuclear constant storage device 3, a three-dimensional core nuclear thermal and hydraulic characteristic analysis device 4, a conversion coefficient storage device 5, a monitoring index analysis device 6, a request An input device 7 and a display device 8 are provided. The plant data input device 2 is connected to a three-dimensional core nuclear thermal and hydraulic characteristic analyzer 4, and the three-dimensional core nuclear thermal and hydraulic property analyzer 4 includes a nuclear constant storage device 3, a monitoring index analyzer 6, a request input device 7 and Each is connected to a display device 8. A conversion coefficient storage device 5 is connected to the monitoring index analysis device 6. The display device 8 is also connected to the monitoring index analysis device 6. The plant data input device 2 is connected to a neutron detector 11 disposed in the core 10 installed in the nuclear reactor 9, and further inputs plant data (core flow rate, core thermal output, control rod position, etc.). To do.

The structure of the control rod that is taken in and out of the core 10 of the nuclear reactor 9 and controls the reactor power will be described with reference to FIG. The control rod 12 has a cross shape in cross section, and has four blades 13 extending in four directions from the axial center. Each blade 13 has a plurality of neutron absorbing rods 18 arranged inside a sheath 14 having a U-shaped cross section. A tie rod 15 is disposed on the axis of the control rod, and a handle 16 is attached to the upper end of the tie rod 15 by welding. A lower support member 17 is attached to the lower end of the tie rod 15 by welding. A speed limiter is provided at the lower end of the lower support member 17. Both end portions of the sheath 14 are welded to the tie rod 15, and the upper end portion of the sheath 14 is attached to the handle 16 by welding. A plurality of neutron absorbing rods 18 filled with B ₄ C, which is a neutron absorbing material, are arranged inside the sheath 14 in each blade 13. The lower end of each neutron absorber rod 18 is supported on the upper surface of the lower support member 17.

  The reactor core performance calculation apparatus of the present embodiment obtains, for example, a monitoring index used for life evaluation of a control rod having the above configuration.

  The nuclear constant storage device 3 stores nuclear constants calculated in advance for each fuel assembly. The three-dimensional core nuclear thermal hydraulic characteristic analysis device 4 reads data from the plant data input device 2 and the nuclear constant storage device 3 and analyzes the output and the like for each node. Specifically, core state data such as core flow rate, core thermal output, control rod position, etc. are read from the plant data input device 2, and the neutron flux distribution calculation in the core is performed using the nuclear constants read from the nuclear constant storage device 3. And thermal hydraulic calculation to calculate power distribution, void fraction distribution, burnup distribution, etc. in the core. In addition, the nuclear constant used for neutron flux distribution calculation is corrected using the measured value of the neutron detector read from the plant data input device 2. The conversion coefficient storage device 5 stores a fitting coefficient of a conversion coefficient for converting from a fuel output calculated in advance for each fuel assembly into a monitoring index of a control rod life. The monitoring index analysis device 6 inputs data from the three-dimensional core nuclear thermal hydraulic characteristic analysis device 4 and the conversion coefficient storage device 5, respectively, and analyzes the monitoring index of the control rod life. The request input device 7 inputs request information from the operator. The display device 8 displays the analysis results in the three-dimensional core nuclear thermal hydraulic characteristic analysis device 4 and the monitoring index analysis device 6. The monitoring index of the control rod life obtained in this embodiment is the neutron absorption amount of the control rod.

Next, each processing in the conversion coefficient storage device 5 and the monitoring index analysis device 6 will be described. The conversion coefficient stored in the conversion coefficient storage device 5 from the fuel output to the monitoring index of the control rod life (in this embodiment, the neutron absorption amount of the control rod) is evaluated in advance using the fuel type, void ratio and burnup as parameters. Keep it. The fuel type, void ratio, and burnup are each information of the fuel type, void ratio, and burnup in the fuel assembly that is loaded in the core and that is adjacent to the control rod for obtaining the monitoring index. In the conversion coefficient storage device 5, the dependence of the conversion coefficient on each parameter is input and stored. The dependence of the conversion factor on each parameter is obtained in advance as follows. Specifically, for each fuel type, the relationship (conversion coefficient) between the fuel output and the neutron absorption amount of the control rod is evaluated in advance for a plurality of burnups with respect to several representative void ratios. Next, for each burnup, fitting is performed using a function (for example, a quadratic function) indicating the relationship between the void ratio and the conversion coefficient. For example, it is assumed that the relationship between the void ratio and the conversion coefficient is fitted with α ₁ = a ₁ x ² + b ₁ x + c ₁ which is a quadratic function. Here, α ₁ is a conversion coefficient, x is a void ratio, and a ₁ , b _1, and c ₁ are fitting coefficients. These fitting coefficients are obtained in advance for each burnup and are stored in the conversion coefficient storage device 5.

  In the monitoring index analysis device 6, data stored in the conversion coefficient storage device 5 (fitting coefficients (a1, b1, c1) given for each fuel type and each burnup), and three-dimensional core nuclear thermal hydraulic characteristics analysis Using the above quadratic function, it is converted for each node by using the fuel output, void ratio and burnup for each node of the fuel assembly adjacent to the control rod for obtaining the monitoring index inputted from the device 4 in the core 10 A coefficient is calculated, and the neutron absorption amount for each node of the control rod adjacent to the fuel assembly is calculated based on the calculated conversion coefficient for each node and the fuel output for each node. Specifically, for each node, using the fitting coefficients for a plurality of burnups input from the conversion coefficient storage device 5, the fitting coefficient corresponding to the burnup input from the three-dimensional core nuclear thermal hydraulic characteristic analyzer 4 is obtained. Interpolate. Using the interpolated fitting coefficient, a conversion coefficient corresponding to the void ratio input from the three-dimensional core nuclear thermal hydraulic characteristic analyzer 4 is calculated. The conversion factor obtained for each node is multiplied by the fuel output of each node input from the three-dimensional core nuclear thermal hydraulic characteristic analyzer 4 to calculate the neutron absorption amount of the control rod for each node. Then, by taking the average of the neutron absorption amounts of the control rods at the four nodes adjacent to the control rod, the neutron absorption amount in the corresponding axial region of the control rod is obtained.

  When the obtained neutron absorption amount of the control rod reaches the limit value of the neutron absorption amount of the control rod corresponding to a preset life, the control rod reaches the end of its life.

  According to this embodiment, the neutron absorption amount of the control rod, which is a monitoring index having a good correlation with the life of the control rod, is changed from the fuel output to the neutron absorption amount of the control rod depending on the fuel type, void ratio, and burnup. By using the conversion coefficient and the fuel output, the neutron absorption amount of the control rod, which is a monitoring index having a better correlation with the control rod life than the conventional thermal neutron irradiation amount, can be obtained with higher accuracy.

  A reactor core performance calculation apparatus according to embodiment 2, which is another embodiment of the present invention, will be described below. The reactor core performance calculation apparatus of the present embodiment is substantially the same as the reactor core performance calculation apparatus 1 of the first embodiment. The monitoring index of the control rod life in this embodiment is the neutron absorption amount of B10 in the control rod. For this reason, in the reactor core performance calculation apparatus of the present embodiment, the data (fitting coefficients (a1, b1, c1) stored in the conversion coefficient storage apparatus 5 are stored in the reactor core performance calculation apparatus 1 of the first embodiment. The value is different from the data (fitting coefficients (a1, b1, c1)) stored in the conversion coefficient storage device 5.

In the three-dimensional core nuclear thermal-hydraulic characteristic analyzer 4, processing similar to that in the first embodiment is performed. Each processing in the conversion coefficient storage device 5 and the monitoring index analysis device 6 in the present embodiment will be described. The conversion coefficient stored in the conversion coefficient storage device 5 from the fuel output to the monitoring index of the control rod life (in this embodiment, the neutron absorption amount of B10 in the control rod) is determined in advance by the fuel type, void ratio and burnup. Evaluate as a parameter. In the conversion coefficient storage device 5, the dependence of the conversion coefficient on each parameter is input and stored. The dependence of the conversion factor on each parameter is obtained in advance as follows. Specifically, for each fuel type, the relationship (conversion factor) between the fuel output and the neutron absorption amount of B10 in the control rod is calculated in advance for a plurality of burnups with respect to several typical void ratios. evaluate. Next, for each burnup, fitting is performed using a function (for example, a quadratic function) indicating the relationship between the void ratio and the conversion coefficient. For example, it is assumed that the relationship between the void ratio and the conversion coefficient is fitted with α ₁ = a ₁ x ² + b ₁ x + c ₁ which is a quadratic function. Here, α ₁ is a conversion coefficient, x is a void ratio, and a ₁ , b _1, and c ₁ are fitting coefficients. These fitting coefficients are obtained in advance for each burnup and are stored in the conversion coefficient storage device 5.

  In the monitoring index analysis device 6, the data stored in the conversion coefficient storage device 5 (fitting coefficients (a 1, b 1, c 1) and the control rod for obtaining the monitoring index input from the three-dimensional core nuclear thermal hydraulic characteristic analysis device 4 The conversion factor is calculated for each node by the above quadratic function using the fuel output, void ratio, and burnup for each node of the fuel assembly adjacent in the reactor core 10, and each of these calculated nodes The neutron absorption amount of B10 for each node in the control rod adjacent to the fuel assembly is calculated based on the conversion coefficient for each node and the fuel output for each node. Is used to interpolate the fitting coefficient corresponding to the burnup input from the three-dimensional core nuclear thermal hydraulic characteristic analyzer 4. Using the inserted fitting coefficient, a conversion coefficient corresponding to the void ratio input from the three-dimensional core nuclear thermal hydraulic characteristic analysis device 4 is calculated, and the conversion coefficient obtained for each node is calculated as the three-dimensional core nuclear thermal hydraulic power. The neutron absorption amount of B10 in the control rod for each node is calculated by multiplying the fuel output of each node inputted from the characteristic analyzer 4. Then, the B10 in the control rod at the four nodes adjacent to this control rod is calculated. By taking the average of the neutron absorption, the neutron absorption of B10 in the corresponding axial region of the control rod is determined.

  When the obtained neutron absorption amount of B10 in the control rod reaches the limit value of the neutron absorption amount of B10 in the control rod corresponding to the preset life, the control rod reaches the end of its life.

  According to the present embodiment, the neutron absorption amount of B10 in the control rod, which is a monitoring index having a good correlation with the life of the control rod, is changed from the fuel output to the B10 in the control rod depending on the fuel type, void ratio, and burnup. The neutron absorption amount of B10 in the control rod, which is a monitoring index having a better correlation with the control rod life than the conventional thermal neutron irradiation amount, is obtained by using the conversion factor to the neutron absorption amount and the fuel output. It can be determined with higher accuracy.

  A reactor core performance calculation apparatus according to embodiment 3, which is another embodiment of the present invention, will be described below. The reactor core performance calculation apparatus according to the present embodiment is substantially the same as the reactor core performance calculation apparatus according to the second embodiment. The monitoring index of the control rod life in this embodiment is the control rod average B10 consumption rate. In the reactor core performance calculation apparatus of the present embodiment, the data stored in the conversion coefficient storage device 5 is the same as the data stored in the conversion coefficient storage device 5 in the reactor core performance calculation apparatus of the second embodiment. However, the processing executed by the monitoring index analysis device 6 is different from the processing executed by the monitoring index analysis device 6 of the reactor core performance calculation apparatus of the second embodiment.

In the three-dimensional core nuclear thermal-hydraulic characteristic analyzer 4, processing similar to that in the first embodiment is performed. Each processing in the conversion coefficient storage device 5 and the monitoring index analysis device 6 in the present embodiment will be described. The conversion coefficient stored in the conversion coefficient storage device 5 from the fuel output to the neutron absorption amount of B10 in the control rod is evaluated in advance using the fuel type, void ratio and burnup as parameters. In the conversion coefficient storage device 5, the dependence of the conversion coefficient on each parameter is input and stored. The dependence of the conversion factor on each parameter is obtained in advance as follows. Specifically, for each fuel type, the relationship (conversion factor) between the fuel output and the neutron absorption amount of B10 in the control rod is calculated in advance for a plurality of burnups with respect to several typical void ratios. evaluate. Next, for each burnup, fitting is performed using a function (for example, a quadratic function) indicating the relationship between the void ratio and the conversion coefficient. For example, it is assumed that the relationship between the void ratio and the conversion coefficient is fitted with α ₁ = a ₁ x ² + b ₁ x + c ₁ which is a quadratic function. Here, α ₁ is a conversion coefficient, x is a void ratio, and a ₁ , b _1, and c ₁ are fitting coefficients. These fitting coefficients are obtained in advance for each burnup and are stored in the conversion coefficient storage device 5.

  In the monitoring index analysis device 6, the data stored in the conversion coefficient storage device 5 (fitting coefficients (a 1, b 1, c 1) and the control rod for obtaining the monitoring index input from the three-dimensional core nuclear thermal hydraulic characteristic analysis device 4 The conversion factor is calculated for each node by the above quadratic function using the fuel output, void ratio, and burnup for each node of the fuel assembly adjacent in the reactor core 10, and each of these calculated nodes The neutron absorption amount of B10 for each node in the control rod adjacent to the fuel assembly is calculated based on the conversion coefficient for each node and the fuel output for each node. Is used to interpolate the fitting coefficient corresponding to the burnup input from the three-dimensional core nuclear thermal hydraulic characteristic analyzer 4. Using the inserted fitting coefficient, a conversion coefficient corresponding to the void ratio input from the three-dimensional core nuclear thermal hydraulic characteristic analysis device 4 is calculated, and the conversion coefficient obtained for each node is calculated as the three-dimensional core nuclear thermal hydraulic power. The neutron absorption amount of B10 in the control rod for each node is calculated by multiplying the fuel output of each node inputted from the characteristic analyzer 4. Then, the B10 in the control rod at the four nodes adjacent to this control rod is calculated. By taking the average of the neutron absorption amount, the neutron absorption amount of B10 in the corresponding axial region of the control rod is obtained, and the number density of B10 that has absorbed neutrons is obtained based on the obtained neutron absorption amount of B10. The average B10 consumption rate of the control rod is obtained by calculating the ratio of the number density of B10 that has absorbed the obtained neutron to the initial number density of B10.

  When the obtained average B10 consumption rate of control rods reaches the limit value of the average B10 consumption rate of control rods corresponding to a preset life, the control rods reach the end of their life.

  According to this embodiment, the control rod average B10 consumption rate, which is a monitoring index having the same definition as the control rod life, depends on the fuel type, void ratio, and burnup, and the neutron absorption amount of B10 in the control rod from the fuel output. The control rod average B10 consumption rate, which is a monitoring index having the same definition as the control rod life, has a better correlation with the control rod life than the conventional thermal neutron irradiation amount, by using the conversion factor and the fuel output. Can be obtained with higher accuracy.

  DESCRIPTION OF SYMBOLS 1 ... Reactor core performance calculation apparatus, 2 ... Plant data input apparatus, 3 ... Nuclear constant memory | storage device, 4 ... Three-dimensional core nuclear thermal-hydraulic characteristic analysis apparatus, 5 ... Conversion coefficient memory | storage device, 6 ... Monitoring index analysis apparatus, DESCRIPTION OF SYMBOLS 9 ... Reactor, 10 ... Core, 12: Control rod, 13 ... Blade, 14 ... Sheath, 15 ... Tie rod, 16 ... Handle, 17 ... Lower support member, 18 ... Neutron absorber rod

Claims (1)

A three-dimensional core nuclear thermal hydraulic analysis device for determining the fuel output, void ratio, and burnup of the fuel assembly adjacent to the control rod for obtaining the monitoring index, and the fuel type, the void ratio, and the burnup as parameters. a conversion factor storage device for storing the fitting coefficients of the conversion factor to the monitoring indicator of the control rod lifetime a conversion factor from the fuel output, the fitting coefficients to be stored in the conversion factor storage device, and the three-dimensional A conversion coefficient is calculated using the fuel output, the void ratio, and the burnup of the fuel assembly input from the core nuclear thermal hydraulic characteristic analysis device , and the control rod is calculated based on the conversion coefficient and the fuel output. as a monitoring indicator of life, the neutron absorption of the control rod, monitoring indicator analysis for determining any one of a neutron absorption and control rod average B10 wear rate of B10 in the control rod Reactor core performance calculation apparatus characterized by comprising a location.

Images